1. Field of the Invention
The present invention relates generally to nuclear reactor internals and more specifically to neutron shields for shielding nuclear reactor pressure vessels from overexposure to neutron flux.
2. Description of Related Art
The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated from and in heat exchange relationship with a secondary side for the production of useful energy. The primary side comprises the reactor pressure vessel enclosing a core internals structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor pressure vessel independently. Each of the parts of the primary side comprising the steam generator, a pump and a system of pipes which are connected to the vessel, form a loop of the primary side. The primary side is also connected to auxiliary circuits, including a circuit for volumetric and chemical monitoring of the pressurized water. The auxiliary circuit, which is arranged branching from the primary circuit, makes it possible to maintain the quantity of water in the primary circuit by replenishing, when required, with measured quantities of water, and to monitor the chemical properties of the coolant water, particularly its content of boric acid, which is important to the operation of the reactor.
The internal components of the reactor pressure vessel typically include upper and lower internals. The upper internals include control rod guide tube assemblies, support columns, conduits for instrumentation which enter the reactor vessel through the closure head, and a fuel assembly alignment structure, referred to as the upper core plate. The lower internals include a core support structure referred to as the core barrel, a core shroud that sits inside the core barrel and converts the circular interior of the core barrel to a stepped pattern that substantially corresponds to the perimeter profile of the fuel assemblies that constitute the core supported between a lower core support plate and the upper core plate. The stepped pattern of fuel assemblies places the fuel assemblies closest to the core barrel at the four cardinal coordinates. As an alternate to the shroud, a bolted baffle former structure consisting of machined horizontal former and vertical baffle plates, has been employed. Whether a shroud or a bolted baffle former structure is employed the core pattern is still generally the same with the fuel assemblies being positioned closest to the core barrel at the cardinal coordinates.
In pressured water moderated and boiling water nuclear reactors, both of which are used for producing steam to drive a steam turbine, fission takes place within the reactor core whereby the energy of the fission products is transmitted to the water. In a pressurized water-moderated reactor the heated water is pumped from the reactor vessel through a heat exchanger in which heat energy is transferred to another circuit of water to form steam to drive a turbine. In a boiling water reactor the energy of the fission products is transmitted to the water in the reactor vessel to form steam to drive the turbine. In both types of reactor it has been found that neutron flux imposed on the vessel from the reactor core results in the material of the vessel becoming embrittled. That is, with prolonged exposure to high energy neutrons the fracture toughness of the vessel is reduced and could ultimately result in fracturing of the vessel should the condition be allowed to continue.
Several design techniques have been utilized in the past to reduce the radiation exposure of the vessel. For example, in pressured water reactors manufacturers have employed a steel thermal shield to reduce the radiation exposure of the vessel. In one such design illustrated in FIG. 1, the core barrel 32, which surrounds and supports the reactor core 14 within the vessel 10, is itself surrounded by a cylindrical steel wall 15 which is two to three inches (5.08 to 7.62 centimeters) thick and is situated approximately an equal distance from the inner surface of the pressure vessel 10 and the outer surface of the core barrel 32. In an alternative design, disclosed in U.S. Pat. No. 3,868,302, recognizing that the neutron flux level impinging on the inner surface of the reactor vessel 10 varies marketably in the circumferential direction, because some of the fuel assemblies 22 within the core 14 are closer to the core barrel 32 than other fuel assemblies 22, the thickness of the core barrel is selectively increased in those regions where a high flux exists. If the thickness of the core barrel is increased in the high flux regions by an amount corresponding to the thickness of the cylindrical thermal shield 15, the maximum radiation exposure of the reactor vessel 10 is essentially the same as the level that would result with a separate cylindrical thermal shield 15.
Another alternative that has been employed is to affix steel plates to the outside of the core barrel in the regions of the cardinal coordinates to essentially achieve the same affect that would be achieved by increasing the thickness of the core barrel in those areas, but at a lower expense. As the size of the reactors increase, requiring larger thermal shields, the steel plates forming the shields can experience large temperature excursions as they heat up and differences in thermal expansion between the shields and the core barrel can strain the means employed to attach the neutron shields to the core barrel. In addition, as the neutron shields become larger, they can affect the pressure drop within the downcomer region 31 between the core barrel and the reactor vessel which, in turn, increases the pressure drop across the reactor vessel and the amount of flow turbulence that is generated in the downcomer annulus.
Accordingly, an improved neutron shield design is desired that will reduce any adverse affect on pressure drop within the downcomer, region between the pressure vessel and the core barrel.
Furthermore, such an improved neutron shield design is desired that will minimize temperature differences between the thermal shield and the core barrel and reduce the strain on the means employed to attach the neutron shields to the core barrel.